Search

The optimization of the 9Be target has been carried out to increase the fast neutron flux and is expected to induce an increase in the epithermal neutron flux. Optimization was carried out by interacting of the 9Be target with protons of 5-30 MeV and studying the effect of variations in the thickness and diameter of the target on the production of fast neutrons. Proton energy optimization results show that the interaction of a 30 MeV proton with a 9Be target produces the highest fast neutron flux with an energy above 10 keV. The optimization results of the Be configuration results in 7 cm thick and 5 cm diameter capable of producing the highest fast neutron flux of 1013 cm²/s. The increase in the fast neutron flux produced by the target resulted in an increase in the epithermal neutron flux from 1 x 109 n/cm²/s to 4.26 x 109 n/cm²/s. [ABSTRACT FROM AUTHOR]

Double layer beam shaping assembly (DLBSA) is a system that moderates fast neutrons into epithermal neutrons. The epithermal neutrons that leave the aperture in the DLBSA system are broadened in the space, thereby reducing the intensity and homogeneity of the epithermal neutron beams. Therefore, it is necessary to improve the design. The development of the DLBSA design was carried out using an extension nozzle. The nozzles are designed using materials made in three configurations, namely Ni+LiF load polyethylene, Pb+LiF load polyethylene, and Bi+LiF load polyethylene. The simulation results show that the addition of a nozzle at the tip of the DLBSA can channel the beam more directionally with high intensity. The addition of nozzles with Ni+LiF load PE material produces an epithermal neutron beam that meets the IAEA standards.

The Li(p,n) and Be(p,n) reactions are widely used as accelerator-based neutron sources for boron neutron capture therapy (BNCT). Because the energy of the neutrons produced by these reactions is high for BNCT, a moderator is required to reduce the energy to the epithermal energy region. The neutron yields and energy spectra derived from the Li(p,n) and Be(p,n) reactions vary with the incident proton energy; a high proton energy results in a high neutron yield. However, a high proton energy can also increase the neutron energy, and in this case, a high neutron energy requires a thick moderator, which decreases the epithermal neutron flux. Notably, these tradeoffs are not completely clear. Furthermore, the upper limit of the epithermal neutron energy used in different BCNT facilities is not the same. Therefore, epithermal neutron flux estimation with applying an appropriate moderator is necessary for comparing epithernal neutron flux depending on the proton energy, target type, and epithermal neutron range definition. In this study, GEANT4 simulations were performed to determine effective moderator materials and thicknesses. After applying a moderator, we examined the epithermal neutron flux by changing the conditions that can influence the epithermal neutron flux, such as the proton energy, target species (Li and Be), and different definitions of the epithermal neutron region. The results showed that the epithermal neutron flux increases with the increasing proton energy until a certain level for both Li and Be targets. For proton energies above 3 MeV for the Li target and above 7 MeV for the Be target, the epithermal neutron flux decreases or its increase rate gradually decreases. And, when the upper energy limit of the epithermal neutron range is changed from 10 keV to 20 keV or 40 keV, the epithermal neutron flux is more than 109 cm−2 s−1 at a proton energy of 3 MeV for Li target and 8 MeV for Be target. • Neutron generation yields and energy spectra from Li(p,n) and Be(p,n) reaction were obtained through GEANT4 simulations. • The combination of MgF 2 and TiF 3 showed a suitable moderator for obtaining epithermal neutrons. • We examined the changes in f epi depending on the proton energy, target type(Li and Be), and epithermal neutron range definition. [ABSTRACT FROM AUTHOR]

To moderate fast neutrons produced by accelerator to appropriate therapeutic energies for boron neutron capture therapy (BNCT), beam shaping assembly (BSA) is required definitely. In this work, based on a model of 2.5 MeV/30 mA proton accelerator, the Monte Carlo simulation software MCNPX was employed to design multi-terminal BSAs. All parameters for both the thermal and epithermal neutron beams at the exit ports of the designed BSAs meet the treatment recommendation values proposed by the International Atomic Energy Agency (IAEA). The clinical parameters of the thermal and epithermal neutron beams were also calculated for clinical indication consideration.

• Update measurement of σ 0 and I 0 cross-sections for the 186W(n,γ)187W reaction. • The qualified neutron spectrum and the updated nuclear decay data were applied in this experiment. • The experimental previous values were normalized for comparison with present results. The thermal neutron radiative capture cross-section (σ 0) and the resonance integral (I 0) of the reaction 186W(n,γ)187W were measured by the thin foil activation method using the well thermalized neutron source at the graphite thermal column facility of the Dalat research reactor. The activities of the irradiated target and monitor samples were measured using a high resolution HPGe gamma-ray spectrometer. The recent updated values of gamma-rays intensities and decay half-life of the 187W nucleus were applied in this experiment to substantially improve the cross-section results compared to the previous measurements. The thermal neutron capture cross- section and the resonance integral of the 186W(n,γ)187W reaction were determined relative to the standard reaction 197Au(n,γ)198Au. The values obtained are 31.76 ± 0.63 and 429.2 ± 25.6 barns for σ 0 and I 0 respectively. Our results and previous measured values are compared, and discrepancies are discussed. [ABSTRACT FROM AUTHOR]

The T0 chopper has the ability to remove the fast neutrons and high-energy gamma rays emitted at time zero of each pulse cycle. The multi-physics instrument of the China Spallation Neutron Source requires high epithermal neutron flux to obtain large momentum transfer, and the traditional single T0 chopper is a disadvantage for the transmission of epithermal neutrons in the long focusing guide. To solve this problem, the two counter-rotating T0 chopper design is proposed that can greatly improve the transmission efficiency of epithermal neutrons. The results of the beam experiment show that compared with the single T0 chopper operating at 50 Hz, the two counter-rotating T0 choppers at the same speed significantly improves the transmission efficiency of epithermal neutrons (especially the shorter wavelength neutrons). For example, when the two blades partially overlap, the transmission efficiencies of 0.05 Å, 0.1 Å, and 0.2 Å wavelength neutrons are increased by 9.4, 4.7, and 4.6 times, respectively, which fully proves that the two counter-rotating T0 chopper design is reasonable, and the experimental results are in line with expectations. The design and application of two counter-rotating T0 choppers at the China Spallation Neutron Source is the first example of this in the world and can be applied to neutron spectrometers with high epithermal neutron flux requirements, and can also be extended to fields such as the energy selection of high-energy particles. [ABSTRACT FROM AUTHOR]

Background: In the field of the treatment by Boron Neutron Capture Therapy (BNCT), an optimized neutron system was proposed. This study (simulation) was conducted to optimize the geometry and composition of neutron system and increase the epithermal neutron flux for the treatment of deep tumors is performed. Materials and Methods: A neutron system for BNCT was proposed. The system included 252Cf neutron source, neutron moderator/reflector arrangement, filter and concrete. To capture fast neutrons, different neutron filters Fe, Pb, Ni and PbF2 with various thicknesses were simulated and studied. Li (with 1 mm thick) was used for filtering of thermal neutrons. Bi with thickness of 1 cm was used to minimize the intensity of gamma rays. Monte Carlo simulation code MCNPX 2.4.0 was used for design of the neutron system and calculation of the neutron components at the output port of the system. Results: For different thicknesses of the filters, the fast neutron flux, the epithermal and thermal flux were calculated at the output port of the system. The spatial distribution of the fast neutron flux, the epithermal flux and gamma flux in human head phantom with the presence of 40 ppm of 10B were obtained. The present calculations showed that Pb filter (about 1 cm) at the output port is suitable for fast neutron capture. The thickness of Li filter was determined due to its high absorption cross-section in thermal region. Bi was used as a gamma filter by the reason of it is good for shielding gamma rays, while having high transmission epithermal neutrons. Conclusion: The epithermal neutron flux has enhanced about 38 percent at the output port of the present system, compared with recent system proposed by Ghassoun et al. At 2 cm depth inside the head phantom, the neutron flux reaches a maximum value about . At this depth, the ratio of the thermal neutron flux to the epithermal flux is about three times, that suggests such a neutron system to treat tumors in the proximity of the depth. In the presence of 10B in the brain, at 2 cm depth the neutron absorption takes place more than other areas of the brain, and consequently the thermal flux is depressed uniformly in the head phantom. Due to high LET and RBE of alpha and 7Li particles (obtained by reaction of boron-neutron), the tumor at the mentioned depth is damaged rather than the around.

Abstract The optimization of a Beam Shaping Assembly design for Boron Neutron Capture Therapy (BNCT), based on a commercial available proton accelerator, has been simulated. The primary goal is to improve the flux of the epithermal neutrons, which is the major drawback in all designed BNCT facilities with portable neutron source, satisfying simultaneously all the other relevant parameters. A BNCT facility has been simulated using the MCNP4B Monte Carlo code. A wide set of material were examined as moderators, while three different materials are used as fast neutron, gamma ray and thermal neutrons filters. Fast neutrons from 9Be(p,n)9B reaction are moderated through the 7LiF and TiF 3 moderators and further reduced using 60Ni as fast neutron filter. The undesired gamma rays and thermal neutrons are removed using Bi and Cd filters respectively. The therapeutic efficacy of the proposed facility was calculated through the dosimetric evaluation in a head phantom. According to the results of the calculations the proposed unit satisfies the recommended by IAEA parameters. Only the flux of the epithermal neutrons remains below the recommended threshold, but is significantly improved. [ABSTRACT FROM AUTHOR]

For designing an appropriate neutron beam, the determination of neutron flux at any irradiation facility is an important key factor. Due to the importance of determining the thermal and epithermal neutron fluxes in a typical thermal column of a reactor, a simple and accurate technique is introduced in this study. Absolute thermal and epithermal fluxes were measured experimentally at a certain point using the foil activation method by neutron bombardment of bare and cadmium covered Au foils. The relative neutron fluxes were also derived simply by means of Monte Carlo simulation by accurate modelling of the reactor components. Finally, by normalization of the relative distribution flux with regard to information about the absolute neutron flux, the accurate thermal and epithermal neutron distributions were derived, separately. [ABSTRACT FROM AUTHOR]

Boron neutron capture therapy (BNCT) is a biochemically-targeted type of radiotherapy, selectively delivering localized dose to tumour cells diffused in normal tissue, while minimizing normal tissue toxicity. BNCT is based on thermal neutron capture by stable 10B nuclei resulting in emission of short-ranged alpha particles and recoil 7Li nuclei. The purpose of the current work was to develop and validate a Monte Carlo BNCT beam model and to investigate contribution of individual dose components resulting of neutron interactions. A neutron beam model was developed in Geant4 and validated against published data. The neutron beam spectrum, obtained from literature for a cyclotron-produced beam, was irradiated to a water phantom with boron concentrations of 100 μg/g. The calculated percentage depth dose curves (PDDs) in the phantom were compared with published data to validate the beam model in terms of total and boron depth dose deposition. Subsequently, two sensitivity studies were conducted to quantify the impact of: (1) neutron beam spectrum, and (2) various boron concentrations on the boron dose component. Good agreement was achieved between the calculated and measured neutron beam PDDs (within 1%). The resulting boron depth dose deposition was also in agreement with measured data. The sensitivity study of several boron concentrations showed that the calculated boron dose gradually converged beyond 100 μg/g boron concentration. This results suggest that 100μg/g tumour boron concentration may be optimal and above this value limited increase in boron dose is expected for a given neutron flux. [ABSTRACT FROM AUTHOR]

Nigerian Research Reactor-1 (NIRR-1) undergo different installation such as permanent Cadmium line due to high demand for epithermal neutrons activation analysis (ENAA) irradiation by the clients and cadmium poison to regulate the reactivity of the reactor with a range value 4.0mk to 3.7mk. Safety and flux stability. Simulations were carried out via Monte Carlo Transport Code MCNP5 for these installations but very few experimental calculation were done. The results of all these research work revealed that the effect of cadmium-line on safety and flux stability is very small. But in the case of cadmium poison, the reactivity of the reactor after 10 years of operation was observe to be decreasing. The results obtained for excess core reactivity after (2.96 and 2.95) mk, and predicted power of 14.65 kW for the two experiment with coolant temperature (12.10 and 12.08) °C respectively. These show that the change in power of the reactor is very small. The outcome of the research will help sample handling capabilities of NIRR-1 and provide useful data to the MNSR. [ABSTRACT FROM AUTHOR]